ML22067A203
| ML22067A203 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/08/2022 |
| From: | Robert Kuntz Plant Licensing Branch III |
| To: | Loeffler R Northern States Power Company, Minnesota |
| Kuntz R | |
| References | |
| EPID L-2021-LLA-0144 | |
| Download: ML22067A203 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Mr. Loeffler, The NRC staff has determined that an audit is necessary to support its review of the Monticello Nuclear Generating Plant, license amendment request to adopt advanced Framatome methodologies. The following is the plan for the audit. If Xcel Energy has any questions or requires information contact me.
Robert Kuntz Senior Project Manager NRC/NRR/DORL/LPL3 This document contains proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390.
Proprietary information is identified by bolded text enclosed within double brackets as shown here (( )).
REGULATORY AUDIT PLAN TO SUPPORT REVIEW OF THE LICENSE AMENDMENT REQUEST REGARDING APPLICATION OF FRAMATOME METHODOLOGIES TO SUPPORT TRANSITION TO ATRIUM 11 FUEL MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 EPID L-2021-LLA-0144
1.0 BACKGROUND
By letter dated July 29, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21211A594), Xcel Energy (the licensee) submitted a license amendment request (LAR) for Monticello Nuclear Generating Plant (MNGP) to allow application of the Framatome analysis methodologies necessary to support a planned transition to ATRIUM 11 fuel under the currently licensed Extended Flow Window (EFW) operating domain.
The NRC staff has determined that a regulatory audit of the background technical analysis for this LAR should be conducted in accordance with the Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-111, Regulatory Audits, for the NRC staff to gain a better understanding of the licensees calculations and other aspects of the LAR.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.0 REGULATORY AUDIT BASIS A regulatory audit is a planned, license or regulation-related activity that includes the examination and evaluation of information that is not on the docket. The basis for the review is the LAR and the NRCs Standard Review Plan Chapter 15, Transient and Accident Analysis.
The staff has determined an audit will be the most efficient path to a timely resolution of issues associated with this LAR review. It is intended to avoid unnecessary burden on the licensee by preventing the generation of requests for additional information (RAIs) when that information is not needed to make a safety determination. Furthermore, interaction to clarify what information is needed should avoid RAI responses that do not fulfill the staffs needs. In this way, followup requests should be minimized.
3.0 REGULATORY AUDIT SCOPE The scope of the audit is the documentation and supporting analysis related to the LAR.
4.0 INFORMATION NEEDS Audit items are included in Appendix A to aid in discussion.
As necessary, the licensee should make available technical staff or contractors who are familiar with the information to assist the NRC staff during the audit.
Additional information needs identified during the audit will be communicated to the designated point of contact.
5.0 TEAM ASSIGNMENTS The NRC staff performing this audit will include, but may not be limited to, the following staff:
Audit Team Diana Woodyatt, Technical Reviewer, Team Lead Ahsan Sallman, Technical Reviewer Mathew Panicker, Technical Reviewer Santosh Bhatt, Technical Reviewer Adam Rau, Technical Reviewer Robert Kuntz, Project Manager 6.0 LOGISTICS The audit will be conducted from January 27, 2022 to February 25, 2022 through remote online access (electronic portal, ePortal, electronic reading room) established by NSPM. A session for OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION presentations, review of audit questions, is planned for the week of February 7, 2022. This session is expected to last three days, but it will be extended if necessary.
7.0 DELIVERABLES At the conclusion of the audit, the NRC staff will conduct an exit briefing and provide a summary of audit results. The NRC staff plans to prepare a regulatory audit summary within 90 days of the completion of the audit.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION APPENDIX A: AUDIT ITEMS FOR REVIEW AND DISCUSSION 1.0 DOCUMENTS TO BE PLACED IN THE ONLINE PORTAL The following documents should be included in the online portal for the NRC staff to review:
1 ANP-3882P Section 3.1 states that The mechanical design of fuel assemblies shall be compatible with co-resident fuel and the reactor core internals.
Provide calculations that show the compatibility of ATRIUM 11 with coresident fuel and the reactor core internals.
2 Provide documentation with description, and results from testing performed to qualify the mechanical design that evaluate the assembly characteristics including fuel assembly axial load structural strength, fuel assembly fretting, fuel assembly static lateral deflection, fuel assembly lateral vibration, fuel assembly impact, spacer grid lateral strength, and tie plate lateral strength described in ANP-3882P.
3 Provide calculation details for how the criterion for fuel bowing, ((
)) is satisfied by Framatome analysis described in ANP-3882P.
4 Provide calculation details of analyses performed for thermal-hydraulic compatibility that demonstrates ATRIUM 11 to be compatible with ATRIUM 10XM fuel that shows ATRIUM 11 assembly flows are within 5% of ATRIUM 10XM assembly flows at rated conditions and within 6% at off-rated conditions described in ANP-3893P.
5 Provide calculations that used best-estimate code, RODEX4 to obtain (1) Power measurement and operational uncertainties, (b) Manufacturing uncertainties, and (3) model uncertainties from ANP-3903P.
6 Provide documentation for thermal margin performance analysis as summarized in Section 3.3 of ANP-3893P.
7 Provide calculation notebook for fuel-specific seismic/LOCA liftoff analysis for (1) with ATRIUM 11 core and (2) mixed core with ATRIUM 11 and ATRIUM 10XM.
8 According to Section 5.1 of ANP-3924P, the ((
)) void quality correlation was developed by Framatome against both the FRIGG void measurements, ATRIUM 10, and ATRIUM 10XM fuel design, however it is applied to ATRIUM 11 fuel design which has a different geometric configuration from ATRIUM 10XM and ATRIUM 10 fuel design.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION a)
Provide documents that can justify how this correlation can be applied to ATRIUM 11 fuel designs.
b)
Provide documents with calculations that justify the use of ((
)) for use with ATRIUM 11 fuel design.
c)
Even though the ATRIUM 11 ((
))
void fraction measurements, S-RELAP5 has been used against previous measurements. Provide calculations that will justify the use of 2-Sigma error of
((
)) is justified for ATRIUM 11 fuel design.
9 In ANP-3925, Appendix A, limitation and condition 26 states:
Plant-specific licensing applications referencing the AURORA-B LOCA evaluation model shall confirm that ((
))
Make available to NRC staff the most recent version of the modeling guideline document FS1-0034030, "Guidelines for Input Development and Problem Execution for AURORA-B LOCA."
10 Make available related documents for NRC staff to review and confirm that the limitations and conditions 10, 14, 15, 20, 23, and 24 in ANP-3934, Appendix A are satisfied.
2.0 AUDIT QUESTIONS NSPM should be prepared to clarify the following during interaction between NSPM and NRC staff.
1 In ANP-3925, Table 3.1, for the event in Updated Safety Analysis Report (USAR) 14A, Section 5.0, Pneumatic System Degradation (Turbine Trip with Bypass and degraded scram speed), the disposition is to address for initial reload with the comment that the results of this event analysis for this event are expected to be bounded by other event (i.e., FWCF) and the event will be analyzed for the initial reload of ATRIUM 11 fuel to confirm the results remain bounded. In case the expectation is not met, explain the process which confirms that this event will be analyzed for each reload.
2 In ANP-3925, Table 3.1, for the event in USAR 14A, Section 5.0, Loss of Stator Cooling, the disposition is to address for initial reload with the comment that the results of this event analysis are expected to be non-limiting and the event will be analyzed for the initial reload of ATRIUM 11 fuel to confirm the results remain non-limiting. In case the OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION expectation is not met, explain the process which confirms that this event will be analyzed for each reload.
3 In ANP-3925, Table 3.1, for the event in USAR Section 14.A, Table 5.1, Inadvertent High Pressure Coolant Injection the disposition status is to address at each reload. As stated in limitation and condition 13, while licensing applications may rely on nominal calculations with the AURORA-B EM for this event in the course of demonstrating that all regulatory limits are satisfied with significant margin, the existing uncertainty methodology may not be applied directly to this event. Explain the method of uncertainty analysis for the CPR in the reload cycle analysis.
4 In ANP-3925, Appendix A, limitation and condition 26 states:
AREVA must continue to use existing regulatory processes for any code modifications made in the ((
)) areas discussed in Section 4 of the SE [safety evaluation report for ANP-10300P-A, Revision 1.]
To confirm adherence to the regulatory requirements, confirm that the Framatome software development procedures are consistent with the licensees existing regulatory processes for code modification made in ((
))
5 In ANP-3934, Section 4.4, ECCS Parameters, the fourth paragraph states that In the Framatome LOCA analysis model, ECCS initiation is assumed to occur when the water level drops to the applicable level setpoint. Provide the description of the applicable water level setpoint for ECCS initiation used in the LOCA analysis, and justify it is conservative.
6 ANP-3934, Section 8.0, states that For non-recirculation line breaks, the core can be reflooded to the top of the active fuel and be adequately cooled indefinitely. Provide a basis for the above statement.
7 ANP-3934, Section 8.0, states:
For recirculation line breaks, the core will initially remain covered following reflood due to the static head provided by the water filling the jet pumps to a level of approximately two-thirds core height. Eventually, the heat flux in the core will not be adequate to maintain a two-phase water level over the entire length of the core. Beyond this time, the upper third of the core will remain wetted and adequately cooled by core spray.
Provide a basis for the above statement.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 8
Referring to ANP-3934, a)
On page vi, under nomenclature, MWR is the abbreviation for metal-water reaction. On page 2-1, the table shows the Local cladding oxidation (max %), result as 11.22. Describe the relationship between MWR, local cladding oxidation (max %),
and the parameter required by 10 CFR 50.46(b)(2).
b)
On page vi, under nomenclature, CMWR is the abbreviation for core average metal-water reaction. On page 2-1, the table shows the Total hydrogen generated (% of total hydrogen possible), result as < 0.97. Describe the relationship between CMWR, total hydrogen generated (% of total hydrogen possible),
and the parameter required by 10 CFR 50.46(b)(3).
9 In ANP-3934, Appendix A, limitation and condition 11 states:
Plant-specific licensing applications referencing the AURORA-B LOCA evaluation model ((
))
The disposition in the amendment request is as follows:
BWR fuel rods are ((
)).
a)
Provide reasons for ((
)) in the above disposition b)
What is the basis for ((
))
10 Describe the process undertaken to meet limitation and condition 27 of ANP-10333P-A, as discussed in Section 8.0 of ANP-3929P.
11 Discuss the ((
)) as described in Section 6.4 of ANP-3924.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 12 ANP-3934, Section 7.1 states that ((
)). What is the value used in the SLO analysis? Describe the ((
].
13 ANP-3934, Table 4.5 provides the following information for the Low-Pressure Injection parameters:
Total system delay from initiating signal until the system is ready to inject is 53.2 seconds and LPCI injection valve stroke time is 35 seconds In ANP-3934, Table 6.2, the single failure case SF-LPCI, implies only 1 LPCI loop flow is used in the analysis.
During normal operation, in the scenario in which the RHR system is placed in the suppression pool cooling mode, or flow test mode, the motor-operated valve, MO-2009, in the suppression pool return line would be open.
(a)
What is the automatic closing time of the suppression pool return valve, MO-2009, on receiving a LOCA signal?
While RHR system is in suppression pool cooling or flow test mode, the return valve, MO-2009, should automatically close on receiving a LOCA signal in the closing time in response to question (a), and the LPCI injection valve must fully open in 35 seconds from receiving LOCA signal. In case the closing time of the return valve MO-2009 is greater than 35 seconds, some of the LPCI flow will be bypassed to the suppression pool and the reactor may not receive the full flow used in the analysis.
(b)
Does the SF-LPCI single failure LOCA analysis consider the above scenario in which the return valve MO-2009 is partially open for a short period of time (the difference after the time the injection valve fully opens and the time the return valve fully closes)? It may be noted that the peak PCT may occur during early short time during a LOCA.
14 10 CFR 50.46(b)(4) requires the plant response to LOCA meets the following acceptance criteria: Calculated changes in core geometry shall be such that the core remains amenable to cooling.
In a seismic event, how is the above criteria met during normal plant operation and during a simultaneous LOCA event?
15 Referring to the 2 CRs presented during the audit: CR 2021-2756 and CR 2022-0152.
(a)
Will the LAR be supplemented based on these CRs, and if so, which of the reports in the LAR will be revised.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION (b)
In CR 2022-0152, the second bullet mentions the parameter fuel rod thermal time constant. Provide a definition of this parameter.
16 ANP-3934, Section 5.3.4 describes the low-pressure core spray (LPCS) line break evaluation as follows:
A break in the LPCS line is expected to have many characteristics similar to ((
)) However, some characteristics of the LPCS line break are unique and are not addressed in other LOCA analyses. Two important differences from other LOCA analyses are that the break flow will exit from the region inside the core shroud and the break will disable one LPCS system. The LPCS line break is assumed to occur just outside the reactor vessel. ((
))
(a)
Explain what are the characteristics that are similar to ((
))
(b)
What are the LPCS line break characteristics that are unique and not addressed in other LOCA analysis, and what are other LOCA analysis.
(c)
If the break flow exits from the region inside the shroud, how will it disable one LPCS system?